For thirty years the Purex process has been the primary method for the recovery of uranium and plutonium from spent nuclear fuel. Despite world-wide use during such time, no practical method has yet been demonstrated for the removal of diluent degradation products. As a consequence, all of the existing Purex facilities constantly encounter problems in emulsification, improper phase disengagements and highly radioactive solvents.
The name "Purex" was derived from Plutonium Uranium Recovery by EXtraction. The preferred extractant is tri-n-butyl phosphate (TBP), which has good radiolytic and chemical stability, and low aqueous solubility. The chelating properties of TBP make it possible not only to efficiently eliminate the undesired fission products and other actinide byproducts, but also to cleanly separate the two desired products, namely, uranium and plutonium, from each other. However, the density of TBP (0.98) is so close to water that dilution in a non-polar low-density organic liquid is necessary to achieve adequate phase separation. Therefore, the Purex process is generally defined as a counter-current solvent extraction process in which a solution of tri-n-butyl phosphate in a non-polar organic diluent is used to recover and purify plutonium and uranium from dissolved, irradiated nuclear fuels or targets. The process has many variations. The percentage of TBP in the solution has been varied in different applications from 3.5 to 50 volume percent, with 30 volume percent being most common. Several diluents have been used, including refined kerosenes and carbon tetrachloride, with the current diluent of choice being n-paraffin oil (for example, predominately n-tridecane). The Purex process has been used to process fuels and targets varying in .sup.235 U enrichment from 0.2 to over 93 percent, to process highly irradiated plutonium, recovering not only the plutonium, but also a mixture of actinides of higher atomic numbers, to separate and purify neptunium-237 byproduct, and to recover .sup.233 U and thorium from irradiated thorium targets.
Typical feed for the Purex process contains uranyl nitrate, nitric acid, plutonium(IV) nitrate, fission products and small concentrations of actinide byproducts. The first cycle of solvent extraction provides the initial decontamination from fission products and, in most cases, the separation of plutonium and uranium. In the first bank of the separators, the plutonium and uranium are extracted into the extraction solution as Pu(NO.sub.3).sub.4 (TBP).sub.2 and UO.sub.2 (NO.sub.3).sub.2 (TBP).sub.2. Neptunium byproduct can be rejected to the aqueous waste stream, if that is desired, by feeding a small stream of nitrous acid or nitrite salt solution into the bank to keep the neptunium in the inextractable Np(V) valence. It is also possible to reject the plutonium to aqueous stream as Pu(III) while extracting Np(IV) as Np(NO.sub.3).sub.4 (TBP).sub.2 by adding a suitable reductant, such as ferrous sulfamate, to the aqueous scrub stream and/or feed. The extraction solution from the first bank contains the desired products and becomes the feed for the second bank. In the second bank of separators, the uranium-plutonium separation is accomplished by reducing plutonium to the relatively inextractable Pu(III) valence, thus stripping it from the extraction solution into the aqueous stream. In the third bank of the first cycle, the uranium is stripped from the extraction solution into a low acid stream.
Four streams exit the first cycle, namely: (1) the spent extraction solution, which is washed and recycled; (2) the aqueous waste; (3) the aqueous uranium product stream, which is concentrated and then processed through a second cycle of solvent extraction for additional decontamination from plutonium and fission products; and (4) the aqueous plutonium product stream, which is processed through a second cycle of solvent extraction for additional fission product decontamination. Typically the plutonium in the fourth exit stream normally contains less than 50 parts of uranium per million parts of plutonium, and the second plutonium cycle provides no uranium decontamination. The uranium in the third exit stream normally contains less than 10 parts of plutonium per billion parts of uranium, but the second uranium cycle has the routine capability of rejecting plutonium to the aqueous waste stream by reducing plutonium in the second cycle feed to Pu(III) with ferrous sulfamate. The only fission products that are sufficiently extractable to become significant contaminants in product streams are .sup.95 Zr, .sup.103 Ru and .sup.106 Ru.
The fission product .sup.131 I (8.04d half life) is sufficiently volatile so that precautions must be taken to prevent undesirable releases of it to the atmosphere. The problem of volatile radioiodine can be dealt with by cooling the fuel long enough for it to decay to insignificant levels or by trapping the iodine on any of a number of silver zeolites.
The diluent normally used in the Purex process is a n-paraffin hydrocarbon. Also continuous extraction solution wash of alternating carbonate (or caustic) and nitric acid are generally used today for extraction solution recycle. There is still the problem of providing degradation-purification balance in the extraction solution such that the extraction solution quality is at an acceptable steady-state, with regards to product losses, fission product decontamination, stage efficiencies, Pu-U separations, washed solvent activity levels and interfacila "crud" levels. Therefore, it is no longer necessary to periodically to replace the entire process extraction solution inventory to maintain acceptable plant performance.
The specific chemicals, that is, extraction solution degradation products (i.e., products resulting from degradation of the extractant and degradation of the diluent which together make up the extraction solution), that are causing the problems have not yet been definitively defined. Some experiments and analytical investiations have identified certain types of chemicals in degraded extraction solution, namely, nitrate esters, nitroalkanes, organic phosphate polymers, dibutyl prosphoric acid, butanol, carboxylic acids, nitroso compounds, alkenes and carboxylic acid polymers. But most of such work was done in kerosene-type diluent, rather than n-paraffin, and the experiments rarely simulated actual steady-state plant conditions with its repetitive cycles of extraction solution irradiation and extraction solution washing. The strong ligands, whatever they are, that tightly bind ruthenium, zirconium and plutonium, in the extraction solution, have little or no adverse effect on plant operation. The types of chemicals that do affect plant performance are: (a) weak ligands that extract fission products or other undesired elements into the extraction solution in the more acidic extraction sections of the banks, and then release them into the aqueous phase along with the products in the lower acid stripping section, and (b) any chemicals that reduce the surface tension between the extraction solution and aqueous phases thus reducing coalescence rate causing stage-efficiency changes which can have very significant effects on process performance.
In judging extraction solution purification techniques, ligand removal from the extraction solution is the criterion which should be used to determine the success or failure of the method. Radioactivity removal is not a valid criterion. It is the ligand that must be removed and it is possible to remove radioactivity without removing the ligand.
Techniques that have been investigated in the past to improve extraction solution purification have fallen into four general categories:
(1) Aqueous washes, including a variety of acids and bases, reductants and oxidants and some complexants for removal of specific radionuclides.
(2) Inorganic sorbants, including activated charcoal, activated MnO.sub.2, silica gel, proprietary aluminium silicates and natural ores.
(3) Organic ion exchange, cation and anion, macroporous and gel types.
(4) Distillation, including flash and vacuum techniques.
U.S. Pat. No. 2,990,240 disclosed the basic Purex process for the recovery of uranium and plutonium from spent nuclear fuel. Patent '240 and the basic Purex process involve a process for segregating and separately recovering uranium values, from an initial aqueous solution containing the same together with plutonium and fission product values. The initial solution is intimately contacted, while maintaining a substantial concentration of nitric acid in the solution, with an organic extractant. The organic extract contains a minor volumetric proportion of a substantially water-immiscible liquid tri-alkyl phosphate dissolved in a major volumetric proportion of a substantially water-immiscible, liquid, non-polar, saturated organic solvent. The relative amounts of tri-alkyl phosphate and uranium values in the solutions in contact are controlled to achieve a concentration of uranium values in the organic extractant of at least 0.35 mole of uranium per mole of tri-alkyl phosphate. Thereby uranium values are preferentially extracted into the organic phase. The resulting uranium bearing organic extract is separated from the remaining plutonium-containing and fission-product-containing aqueous raffinate. Thereupon, the separated uranium-bearing organic extract is intimately contacted with water, thereby stripping uranium values therefrom. Thereafter the resulting stripped-uranium-containing aqueous phase is separated from the spent organic extractant. The problem is to remove the impurities from the spent organic extractant, which normally still contains radioactive materials.
U.S. Pat. No. 4,059,671 teaches a method for increasing the lifetime of an extraction medium which contains an organophosphorous acid ester and a hydrocarbon, and which is used for reprocessing spent nuclear fuel and/or breeder materials. The extraction medium impurities resulting from chemical and/or radiolytic decomposition and complex compounds of such impurities with radionuclides are removed by washing the extraction medium with an aqueous hydrazine hydrate wash solution. The used extraction medium is brought into intimate contact with an aqueous hydrazine hydrate wash solution having a concentration of between about 0.1 and 1.0 molar, at a temperature between about 20.degree. to about 75.degree. C. Then the aqueous hydrazine hydrate wash solution is separated from the extraction medium.
U.S. Pat. No. 4,404,130 discloses a process for the plutonium decontamination of an organic solvent useful in the reprocessing of irradiated nuclear fuel. The process includes adding a reducing agent to the organic solvent. The reducing agent is soluble in the organic solvent and is a dialkyl dithiophosphoric acid. The organic solvent containing the reducing agent is brought into contact with an acid aqueous solution. The aqueous solution containing the plutonium is separated from the decontaminated organic solvent.
The extraction solution used in nuclear fuels reprocessing plants is normally regenerated with wash solutions containing alkali, such as, sodium carbonate or sodium hydroxide. While these wash solutions do remove the bulk of the impurities, they do not remove diluent degradation products which result in the generation of surfactants and complexants for fission products. Eventually such impurities build up in the extraction solution and render the material unsuitable for its intended use. The extraction solution is then disposed of, usually by burning. Such disposal requires specialized equipment to prevent toxic emmissions to the atmosphere.